Nuclear reactors have their fuel contained in sealed cladding for the isolation of the nuclear fuel from the moderator/coolant system. The term cladding, as used herein, refers to a zirconium based alloy tube. Often the cladding will be composed of layers including a zirconium alloy substrate and an unalloyed zirconium barrier.
The cladding--nominally in the order of 0.030 inches thick--is formed in the shape of a tube with the nuclear fuel contained typically in pellet form therein. These pellets are stacked in contact with one another for almost the entire length of each cladding tube, which cladding tube is in the order of 160 inches in length. Typically, the cladding tube is provided with springs for maintaining the axial position of the fuel pellets and so-called "getters" for absorbing excess moisture and hydrogen. The internal portions of the fuel rod are pressurized with helium to help conduct the heat from the fuel material to the cladding.
Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel cladding material since they have low neutron absorption cross sections and, at temperatures below about 350.degree. C., are strong, ductile, extremely stable and relatively nonreactive in the presence of demineralized water or stem. "Zircaloys" are a family of corrosion-resistant zirconium alloy cladding materials. They are composed of 98-99% by weight zirconium, with the balance being tin, iron, chromium, and nickel. "Zircaloy-2" and "Zircaloy-4" are two widely-used zirconium-based alloys for cladding. Zircaloy-2 has on a weight basis about 1.2 to 1.7 percent tin; 0.13-0.20 percent iron; 0.06-0.15 percent chromium and 0.05 to 0.08 percent nickel. Zircaloy-4 has essentially no nickel and about 0.2% iron but is otherwise substantially similar to Zircaloy-2.
Zircaloy cladding defects may occur due to various causes including debris induced fretting and pellet-cladding interaction. In the first of these, debris lodges next to the cladding and vibrates or frets against the cladding wall under the influence of the passing coolant. Such vibration continues until the cladding wall is penetrated. Pellet-cladding interaction is caused by the interactions between the nuclear fuel, the cladding, and the fission products produced during the nuclear reaction. It has been found that this undesirable effect is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding in coincidence with corrosive fission product species causing stress corrosion cracking in the cladding.
To combat defects due to pellet-cladding interaction, some cladding includes pure zirconium or low alloy content zirconium barrier layers metallurgically bonded to the inner surface of the tubing. The pioneering work on barrier layer cladding is described in U.S. Pat. Nos. 4,200,492 and 4,372,817 to Armijo and Coffin, U.S. Pat. No. 4,610,842 to Vannesjo, and U.S. Pat. No. 4,894,203 to Adamson. Barrier layers have been found to effectively prevent damage to the cladding due to interaction with the pellet. However, if the cladding wall is compromised in some manner (e.g. perforated or split by debris fretting), and water enters the fuel rod interior, the barrier layer can be rapidly oxidized.
To protect the zirconium barrier from such oxidation should a cladding breach occur, a three layer structure may be employed. Such structures include a corrosion resistant inner liner bonded to the fuel side of the barrier. They are described in U.S. patent application Ser. No. 08/091,672 entitled METHOD FOR MAKING FUEL CLADDING HAVING ZIRCONIUM BARRIER LAYERS AND INNER LINERS and U.S. patent application Ser. No. 08/092,188 entitled INNER LINERS FOR FUEL CLADDING HAVING ZIRCONIUM BARRIER LAYERS, both of which were filed on Jul. 14, 1993, assigned to the assignee hereof, and incorporated herein by reference for all purposes. While such linings can protect against rapid oxidation, they may still be susceptible to "secondary defects" in the cladding at locations away from the primary defect (the primary defect being the initial breach in the cladding wall).
After a fuel element has suffered a primary breach, it can sometimes be used for some period of time in a reactor. However, it has been observed that "secondary defects" sometimes occur as a result of coolant entering through the primary breach. Such secondary defects are often much worse than the primary failures, and the secondary defects can allow release of large amounts of fission products and fuel material from erosion/corrosion of the fuel material. Post-mortem studies of fuel rods indicate that the secondary failures are often due to localized hydriding (ZrH.sub.2) of the cladding.
When the fuel rod is initially breached, the coolant water enters the tube and flashes to steam. It is believed that, the zirconium or zirconium based alloy inner surface of the fuel cladding tube reacts with the steam which has entered the rod by the following corrosion reaction: EQU Zr+2H.sub.2 O .fwdarw.ZrO.sub.2 +2[(p)H.sub.2 (abs)+(1-p)H.sub.2 (gas)](1)
In this equation, H.sub.2 (abs) is the portion of the corrosion generated hydrogen that is absorbed by the metal, H.sub.2 (gas) is the portion which is released into the rod atmosphere, and "p" is the fraction of hydrogen picked up (absorbed) by the metal.
Normally, zirconium is covered by a thin protective oxide film that protects against hydride formation. When this oxide film is scratched or otherwise becomes defective, the protective zirconium oxide will reform over the bare zirconium surface. However, when a condition known as "oxygen starvation" occurs within the cladding interior, the oxide film may not form fast enough to prevent a hydride blister from forming.
As corrosion proceeds by the above reaction, an increasing amount of H.sub.2 O in the gas phase is replaced with H.sub.2. The ratio of H.sub.2 to H.sub.2 O in the gas phase increases with distance from the initial defect location as time progresses. This occurs because the hydrogen diffuses along the tight gap between the fuel and the cladding while the steam reacts with the zirconium in the cladding wall. At a point removed from the primary defect, the local gas phase H.sub.2 /H.sub.2 O ratio eventually becomes sufficiently high (e.g., of the order of 1000/1) that oxygen starvation occurs and the inner surface of the cladding tube forms a local, massive zirconium hydride by direct reaction with the hydrogen in the gas phase.
The massive hydrides are sometimes referred to as "sunbursts" or hydride "blisters" because of a microscopic distribution of associated hydrides and the protrusion from the cladding inner surface which results from the larger volume associated with hydrides in comparison to zirconium metal or alloy. These massive hydrides regions are extremely brittle and prone to self-generated cracks. Thus, they can result in the catastrophic secondary defects described above if subjected to a stress, as due to power increase in the fuel rod.
It is apparent from the above that there exists a need for a cladding that resists formation of hydride defects in the event of a cladding breach.